Qucosa – Hemholtz-Zentrum Dresden-Rossendorf
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Thermomechanische Modellierung eines Reaktordruckbehälters in der Spätphase eines Kernschmelzunfalls
Get PDFConsidering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power
Aufbau eines Versuchsplatzes für die Positronen-Emissions-Tomographie
Get PDFDie Positronen-Emissions-Tomographie (PET) hat sich als bildgebendes Verfahren in der klinischen Routine sowie der medizinischen, biologischen und pharmazeutischen Forschung etabliert. Daraus ergibt sich die Notwendigkeit, Mediziner sowie Physiker und Ingenieure mit der speziellen Ausrichtung auf multidisziplinäre Forschungsgebiete der Biologie und Medizin bzw. medizintechnischen Anwendungen im Rahmen ihrer Ausbildung möglichst praxisnah mit dieser Technik vertraut zu machen. Zu diesem Zweck soll ein Versuchsplatz für die PET aufgebaut werden, an welchem die allgemeinen Prinzipien der Computertomographie (CT), die Grundlagen der PET sowie die PET als kernphysikalisches Multiparameter-Messverfahren vermittelt werden. Für die Realisierung dieser Aufgabenstellung gibt es folgende Vorgaben: - Der Tomograph besteht aus zwei kommerziellen, in Koinzidenz betriebenen, ortsempfindlichen Szintillationsdetektoren. - Zum Gewinnen vollständiger Projektionsdatensätze wird ein Computer gesteuertes Bewegungssystem für die Translation und die Rotation verwendet. - Die Signalverarbeitung basiert auf Standardmodulen der kernphysikalischen Messtechnik. - Zur Gewährleistung einer flexiblen, den unterschiedlichen Anforderungen genügenden Auswertung erfolgt die Speicherung der Daten im Listen-Modus. - Die tomographische Rekonstruktion mittels gefilterter Rückprojektion ist in die laufende Messung (Online-Betrieb) integriert. Daneben besteht die Möglichkeit, die Daten auch nach der Messung in unterschiedlichen Konfigurationen zu rekonstruieren (Offline-Betrieb). Diese Diplomarbeit soll zudem als Grundlage für das zu erstellende Lehrmaterial fungieren. Dabei wird zuerst auf die Anwendung und Entwicklung der PET eingegangen (Abschnitt 1.1). Im zweiten Kapitel werden die für die PET notwendigen Grundbegriffe erklärt sowie die physikalischen und mathematischen Prinzipien und Wirkungsweisen dargestellt. Daran anschließend wird im dritten Kapitel der Aufbau des PET-Versuchsplatzes beschrieben, wobei auch näher auf die einzelnen ausgewählten Komponenten eingegangen wird. Im vierten Kapitel werden systemspezifische Größen für den Versuchsplatz hergeleitet. Darauf folgend wird in Kapitel 5 die Software detaillierter vorgestellt. Die Ergebnisse der durchgeführten Messungen sind im sechsten Kapitel zu finden. Dabei handelt es sich sowohl um Messungen zur Charakterisierung des Messplatzes als auch um vollständige tomographische Rekonstruktionen von Punktquellen. Abschließend werden die Besonderheiten und Merkmale des aufgebauten Tomographen nochmals zusammengefasst sowie ein Ausblick gegeben (Kapitel 7)
The European project FLOMIX-R: Description of the experimental and numerical studies of flow distribution in the reactor primary circuit(Final report on WP 3)
Get PDFThe flow distribution in the primary circuit of the pressurized water reactor was studied with experiments and Computational Fluid Dynamics (CFD) simulations. The main focus was on the flow field and mixing in the downcomer of the pressure vessel: how the different factors like the orientation of operating loops, the total loop flow rate and the asymmetry of the loop flow rates affect the outcome. In addition to the flow field studies the overall applicability of CFD methods for primary circuit thermal-hydraulic analysis was evaluated based on the CFD simulations of the mixing experiments of the ROCOM (Rossendorf Coolant Mixing Model) test facility and the mixing experiments of the Paks NPP. The experimental part of the work in work package 3 included series of steady state mixing experiments with the ROCOM test facility and the publication of results of Paks VVER-440 NPP thermal mixing experiments. The ROCOM test facility models a 4-loop KONVOI type reactor. In the steady-state mixing experiments the velocity field in the downcomer was measured using laser Doppler anemometry and the concentration of the tracer solution fed from one loop was measured at the downcomer and at the core inlet plane. The varied parameters were the number and orientation of the operating loops, the total flow rate and the (asymmetric) flow rate of individual loops. The Paks NPP thermal mixing experiments took place during commissioning tests of replaced steam generator safety valves in 1987-1989. It was assumed that in the reactor vessels of Paks VVER-440 NPP equipped with six loops the mixing of the coolant is not ideal. For the realistic determination of the active core inlet temperature field for the transients and accidents associated with different level temperature asymmetry a set of mixing factors were determined. Based on data from the online core monitoring system and a separate mathematical model the mixing factors for loop flows at the core inlet were determined. In the numerical simulation part of the work package 3 the detailed measurements of ROCOM tests were used for the validation of CFD methods for primary circuit studies. The selected steady state mixing experiments were simulated with CFD codes CFX-4, CFX-5 and FLUENT. The velocity field in the downcomer and the mixing of the scalar were compared between CFD simulations and experiments. The CFD simulations of full scale PWR included the simulation of Paks VVER-440 mixing experiment and the simulation of Loviisa VVER-440 downcomer flow field. In the simulations of Paks experiments the experimental and simulated concentration field at the core inlet were compared and conclusions made concerning the results overall and the VVER-440 specific geometry modelling aspects like how to model the perforated elliptic bottom plate and what is the effect of the cold leg bends to the flow field entering to the downcomer. With Loviisa simulations the qualitative comparison was made against the original commissioning experiments but the emphasis was on the CFD method validation and testing. The overall conclusion concerning the CFD modelling of the flow field and mixing in the PWR primary circuit could be that the current computation capacity and physical models also in commercial codes is beginning to be sufficient for simulations giving reliable and useful results for many real primary circuit applications. However the misuse of CFD methods is easy, and the general as well as the nuclear power specific modelling guidelines should be followed when the CFD simulations are made
Beitrag zur Modellierung der Schmelzerückhaltung im RDB nach Verlagerung von Corium in das untere Plenum: Berechnung des Temperaturfeldes und der viskoplastischen Verformung der Behälterwand
Get PDFBezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreak-tor (LWR) ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Es wurden bereits eine Reihe von Experimenten durchge-führt, welche Erkenntnisse hierüber liefern sollen. Vom Institut für Sicherheitsforschung des FZR wurde ein Finite-Elemente-Modell er-stellt, das sowohl die Temperaturfeldberechnung für die Wand als auch die elasto-plastische Mechanik der Behälterwand beschreibt. Dabei wurde ein fortgeschrittenes Modell für das Kriechen und für die Materialschädigung entwickelt und an Hand von experimentellen Daten validiert. Die thermischen und mechanischen Berechnungen sind rekursiv und sequentiell gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Das Modell wurde für Voraus- und Nachrechnungen der FOREVER-Experimente, die den RDB eines LWR im Maßstab 1:10 nachbilden, angewendet. Diese Experimente wurden an der KTH Stockholm durchgeführt. Die Ergebnisse der Berechnungen sind qualitativ und quantitativ sehr zufriedenstellend. Erste Rechnungen für eine LWR-Geometrie wurden durchgeführt, um Unterschiede und Gemeinsamkeiten zwischen prototypischen Szenarien und skalierten Experi-menten herauszuarbeiten
Energiedispersive Untersuchung der Wechselwirkung schneller Neutronen mit Materie - gemeinsamer Abschlussbericht der DFG Projekte GR 1674/2 und FR 575/5
Get PDFSynthesis of silicon nanocrystal memories by sputter deposition
Get PDFAim of this work was, to investigate the preparation of Si NC memories by sputter deposition. The milestones are as follows: - Review of relevant literature. - Development of processes for an ultrathin tunnel-oxide and high quality sputtered SiO2 for use as control-oxide. - Evaluation of methods for the preparation of an oxygen-deficient silicon oxide inter-layer (the precursor of the Si NC layer). - Characterization of deposited films. - Establishment of techniques capable of probing the phase separation of SiOx and the formation of Si NC. - Establishment of annealing conditions compatible with the requirements of current CMOS technology based on experimental results and simulations of Si NC formation. - Preparation Si NC memory capacitors using the developed processes. - Characterization of these devices by suitable techniques. Demonstration of their memory functionality
The European project FLOMIX-R: Description of the slug mixing and buoyancy related experiments at the different test facilities(Final report on WP 2)
Get PDFThe goal of the work described in this report was the experimental investigation of the mixing of coolant with different quality (temperature, boron concentration) in nuclear reactors on the way from the cold leg through the downcomer and lower plenum to the core inlet in a systematic way. The obtained data were used for the clarification of the mixing mechanisms and form a data basis for the validation of computational fluid dynamics (CFD) codes. For these purposes, experiments on slug mixing have been performed at two test facilities, modelling different reactor types in scale 1:5, the Rossendorf and Vattenfall test facilities. The corresponding accident scenario is the start-up of first main coolant pump (MCP) after formation of a slug of lower borated water during the reflux-condenser mode phase of a small break loss of coolant accident (LOCA). The matrices for the experiments were elaborated on the basis of the key phenomena, being responsible for the coolant mixing during pump start-up. Slug mixing tests have also been performed at the VVER-1000 facility of EDO Gidropress to meet the specifics of this reactor type. The mixing of slugs of water of different quality is also very important for pre-stressed thermal shock (PTS) situations. In emergency core cooling (ECC) situations after a LOCA, cold ECC water is injected into the hot water in the cold leg and downcomer. Due to the large temperature differences, thermal shocks are induced at the reactor pressure vessel (RPV) wall. Temperature distributions near the wall and temperature gradients in time are important to be known for the assessment of thermal stresses. One of the important phenomena in connection with PTS is thermal stratification, a flow condition with a vertical temperature profile in a horizontal pipe. Due to the fluctuating character of the flow, this may cause thermal fatigue in the pipe. Besides of thermal fatigue, a single thermal shock can also be relevant for structural integrity, if it is large enough, especially in the case, that the brittle fracture temperature of the RPV material is reduced due to radiation embrittlement. Therefore, additional to the investigations of slug mixing during re-start of coolant circulation, the mixing of slugs or streams of water with higher density with the ambient fluid in the RPV was investigated. The aim of these investigations was to study the process of turbulent mixing under the influence of buoyancy forces caused by the temperature differences. Heat transfer to the wall and thermal conductivity in the wall material have not been considered. Experiments on density driven mixing were carried out at the Rossendorf and the Fortum PTS facilities
DYN3D version 3.2 - code for calculation of transients in light water reactors (LWR) with hexagonal or quadratic fuel elements - description of models and methods -
Get PDFDYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations